IR 05000293/2007005
Download: ML080320209
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 475 ALLENDALE ROAD KING OF PRUSSIA, PA 19406 February 1, 2008
Mr. Kevin Bronson Site Vice President Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360-5508
SUBJECT: PILGRIM NUCLEAR POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000293/2007005
Dear Mr. Bronson:
On December 31, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Pilgrim Nuclear Power Station (PNPS). The enclosed report documents the results, which were discussed on January 9, 2008, with you and members of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
This report documents one self-revealing finding of very low safety significance (Green) for which no violation of NRC requirements was identified.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,/RA/ Raymond J. Powell, Chief Projects Branch 5 Division of Reactor Projects Docket No. 50-293 License No. DPR-35
Enclosure:
Inspection Report 05000293/2007005
w/Attachment:
Supplemental Information
cc w/encl:
SUMMARY OF FINDINGS
..............................................................................................................3
REPORT DETAILS
..........................................................................................................................4
REACTOR SAFETY
........................................................................................................................4 1R01 Adverse Weather Protection...............................................................................4
1R04 Equipment Alignment.........................................................................................
.4
1R05 Fire Protection....................................................................................................
.6
1R06 Flood Protection Measures.................................................................................
.6 1R11 Licensed Operator Requalification......................................................................7 1R12 Maintenance Effectiveness.................................................................................9 1R13 Maintenance Risk Assessments and Emergent Work Control...........................9 1R15 Operability Evaluations......................................................................................10
1R19 Post-Maintenance Testing.................................................................................11 1R20 Refueling and Other Outage Activities..............................................................11 1R22 Surveillance Testing..........................................................................................12
RADIATION SAFETY
....................................................................................................................12 2OS3 Radiation Monitoring Instrumentation and Protective Equipment.....................12 2PS3 Radiological Environmental Monitoring Program and Radioactive Material Control Program.....................................................................................14
OTHER ACTIVITIES
[OA].............................................................................................................14
4OA1 Performance Indicator (PI)................................................................................14
4OA2 Identification and Resolution of Problems.........................................................15 4OA3 Event Follow-up.................................................................................................18
4OA6 Meetings, Including Exit.....................................................................................20
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
......................................................................................................A-1
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED...........................................................A-1
LIST OF DOCUMENTS REVIEWED
..........................................................................................A-1
LIST OF ACRONYMS
.................................................................................................................A-7
- OF [[]]
- FINDIN [[]]
GS
IR 05000293/2007-005; 10/01/2007-12/31/2007; Pilgrim Nuclear Power Station; Event
Follow-up.
The report covered a 13-week period of inspection by resident and region-based inspectors.
One Green finding was identified. The significance of most findings is indicated by their color
(Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance
Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be
assigned a severity level after
NRC's program for overseeing
the safe operation of nuclear power reactors is described in
- NUR [[]]
EG-1649, "Reactor Oversight
Process," Revision 4, dated December 2006.
- A. [[]]
- NRC -Identified and Self-Revealing Findings Cornerstone: Initiating Events Green. A Green self-revealing finding was identified for Entergy=s failure to ensure the proper verification and calibration of vacuum trip switch
RFO) 16. Specifically, personnel did not ensure that the proper verification/calibration
technique was employed to determine the as-found low condenser vacuum turbine trip
setpoint. Additionally, when the technician identified that the as-found data was
significantly outside of historical as-found values, he did not question the validity of the
data nor did he obtain a peer check. The technician then calibrated the instrument using
the incorrect as-found data which resulted in an incorrect low vacuum trip setpoint and a
subsequent turbine trip and reactor scram on July 10, 2007. This finding is more than minor because it is associated with the human performance
attribute of the Initiating Events Cornerstone and affects the cornerstone objective of
limiting the likelihood of those events that upset plant stability during power operations. The finding is of very low safety significance (Green) because it did not contribute to both
the likelihood of a reactor trip and the likelihood that mitigation equipment would be
unavailable. This finding has a cross-cutting aspect in the area of Human Performance,
Work Practices, because Entergy proceeded in the face of uncertainty or unexpected
circumstances when the VTS-1 setpoint was found significantly outside of expected as-
found values. H.4(a) (Section
- REPORT [[]]
DETAILS
Summary of Plant Status
Pilgrim Nuclear Power Station (PNPS) operated at or near 100 percent power during the
inspection period with the following exceptions: On October 30, 2007, Entergy reduced power to
approximately 48 percent to perform a thermal backwash on the main condenser. Entergy
resumed 100 percent power operation on October 31, 2007. On December 10, 2007, Entergy
shut down and commenced a planned outage to repair leaking safety relief valve, RV-203-3B.
Entergy restored the plant to 100 percent power on December 13, 2007. The plant remained at
or near 100 percent for the remainder of the inspection period.
1.
SAFETY Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01)
a. Inspection Scope (3 samples - 1 seasonal readiness, 2 impending adverse weather) The inspectors performed a review of cold weather preparations during the onset of the
cold weather season to evaluate the site's readiness for seasonal susceptibilities. The inspectors reviewed Entergy's preparations for cold weather and its impact on the
protection of safety-related systems, structures and components (SSCs). The inspection
focused on the intake structure, the station blackout diesel generator and the condensate
storage and transfer system. The inspection was intended to ensure that Entergy's
equipment, instrumentation, and supporting structures were configured in accordance
with Entergy's procedures and that adequate controls were in place to ensure
functionality of the systems in cold weather. The inspectors also conducted a site
walkdown on November 1, 2007, to assess Entergy's readiness for the potential affects
of hurricane Noel. The inspectors verified that all outside objects were properly anchored
or tied down. In addition, the inspectors conducted a site walkdown on December
2, 2007, to evaluate site preparations for an approaching coastal storm with
accompanying high winds.
b. Findings No findings of significance were identified.
1R04 Equipment Alignment (71111.04)
.1 Partial System Walkdowns (71111.04Q)
a. Inspection Scope (4 samples)
The inspectors performed four partial system walkdowns during this inspection period.
The inspectors reviewed the documents listed in the Attachment to determine the correct
5system alignment. The inspectors conducted a partial walkdown of each system to determine if the critical portions of the selected systems were correctly aligned in
accordance with these procedures and to identify any discrepancies that may have had
an effect on operability. The walkdowns included selected switch and valve position checks, and verification of electrical power to critical components. Finally, the inspectors
evaluated other elements, such as material condition, housekeeping, and component
labeling. The following systems were reviewed based on their risk significance for the
given plant configuration: $ "B" Reactor Building Closed Cooling Water (RBCCW) system during degradation of the "A"
RBCCW pump following completion of 3.M.3-47.2, "'B' Train Functional Test of Individual Load Shed
Component.@ b. Findings No findings of significance were identified.
.2 Complete System Walkdown (71111.04S)
a. Inspection Scope (1 sample) The inspectors completed a detailed review of the standby gas treatment (SBGT) system
to verify the functional capability of the system. The inspectors conducted a walkdown of the system to verify that the critical components such as valves, switches, and breakers
were aligned in accordance with procedures and to identify any discrepancies that could
have an effect on operability. The inspectors discussed system health with the system engineer and conducted a
review of outstanding maintenance work orders to verify that the deficiencies did not
significantly affect the
- SB [[]]
GT system function. The inspectors also reviewed the
condition report (CR) database to verify that equipment problems were being identified
and appropriately resolved. In addition, the inspectors reviewed recent test results to
ensure the air system leakage and charcoal filter efficiency met the requirements of
Technical Specifications (TS) and procedures. Documents reviewed during the
inspection are listed in the Attachment.
b. Findings No findings of significance were identified.
61R05 Fire Protection (71111.05) Fire Protection - Tours (71111.05Q)
a. Inspection Scope (8 samples)
The inspectors performed walkdowns of eight fire protection areas during the inspection
period. The inspectors reviewed Entergy's fire protection program to determine the
required fire protection design features, fire area boundaries, and combustible loading
requirements for the selected areas. The inspectors walked down these areas to assess
Entergy's control of transient combustible material and ignition sources. In addition, the
inspectors evaluated the material condition and operational status of fire detection and
suppression capabilities, fire barriers, and any related compensatory measures. The inspectors then compared the existing conditions of the areas to the fire protection
program requirements to ensure all program requirements were being met. Documents
reviewed during the inspection are listed in the Attachment. The fire protection areas
reviewed were: $ Fire Zone 5.2, "B" Train Salt Service Water Pump Room; $ Fire Zone 1.22, "B" Reactor Building Closed Cooling Water Pumps and Heat Exchanger Rooms; $ Fire Zone 4.2, "B" Emergency Diesel Day Tank Room; $ Fire Zone 4.4, "A" Emergency Diesel Day Tank Room; $ Fire Area 1.9, Fire Zone 2.2, "A" Switchgear and Load Center Room; $ Fire Area 1.9, Fire Zone 3.5, Vital Motor Generator Set Room; $ Fire Zone 1.3, High Pressure Coolant Injection Pump/Turbine Room; and $ Fire Zone 2.3, Battery Room AA.@ b. Findings No findings of significance were identified.
1R06 Flood Protection Measures (71111.06) Internal Flooding Inspection
a. Inspection Scope (1 sample) The inspectors walked down selected areas of the plant including the cable spreading
room, vital Motor Generator set, and
- HP [[]]
CI pump room to assess the effectiveness of
Entergy's internal flood control measures. The inspectors assessed the condition of
watertight doors, floor sump systems, curbing, hatch and conduit seals, and floor drains.
The inspectors reviewed
- NRC [[]]
IN-2007-01, Recent
Operating Experience Covering Hydrostatic Barriers," to determine whether Entergy was
identifying internal flooding issues and taking appropriate corrective actions. The
references used for this inspection are listed in the Attachment to this report.
b. Findings No Findings of significance were identified.
1R11 Licensed Operator Requalification (71111.11)
.1 Resident Inspector Quarterly Review (71111.11Q)
a. Inspection Scope (1 sample) The inspectors observed licensed operator requalification training on November 6, 2007.
Specifically, the inspectors observed classroom Senior Reactor Operator (SRO) training
on Emergency Planning, Emergency Action Level (EAL) Classification, and Protective
Action Recommendation (PAR) procedures and processes. The inspectors assessed
the training to determine if the training adequately prepared the
classification levels and to conduct PAR assessments. The inspectors reviewed the
applicable training objectives to determine if they had been achieved. The inspectors
verified that issues identified during the classroom session were entered into the
corrective action program. Documents reviewed during the inspection are listed in the
Attachment.
b. Findings No findings of significance were identified.
.2 Licensed Operator Requalification (71111.11B)
a. Inspection Scope (1 sample) The following inspection activities were performed using
- NUR [[]]
EG 1021, Revision 9,
"Operator Licensing Examination Standards for Power Reactors," Inspection Procedure
7111111, "Licensed Operator Requalification Program," Appendix A,
"Checklist for Evaluating Facility Testing Material" and Appendix B, "Suggested Interview Topics." The inspectors reviewed documentation of operating history since the last requalification
program inspection. Documents reviewed included NRC inspection reports and licensee
CRs that involved human performance issues. The purpose of the review was to ensure
operational events that occurred during the last two years were not indicative of possible
training deficiencies. The inspectors also discussed facility operating events with the resident staff. The inspectors reviewed comprehensive written exams (these exams were administered
in the fall, 2006), and the scenarios and job performance measures administered during
the weeks of September 10 and 17, 2007, to ensure the quality of these exams met or
exceeded the criteria established in the Examination Standards and 10 CFR 55.59,
"Requalification." The inspectors observed the administration of the operating exams to
two crews.
8Conformance with simulator requirements specified in 10 CFR 55.46, "Simulation Facilities" The inspectors observed simulator performance during the conduct of the examinations,
and reviewed simulator discrepancy reports to determine whether facility staff was complying with the requirements of 10 CFR 55.46. The inspectors reviewed a sample of
simulator tests including transients; normal and steady state; malfunctions; and core
performance tests. Conformance with operator license conditions The inspectors determined whether the operators were complying with the conditions of
their license by reviewing the following: $ five medical records (The records were complete; restrictions noted by the doctor were reflected on the individual's license; and physical exams were given within
months.); $ eight proficiency watch-standing records and one reactivation record (Records indicated the licensed operators conformed with proficiency and reactivation
watch-standing requirements of 10 CFR 55.53, Conditions of Licenses.); and $ remediation training records for four licensed operators (These operators had failed either an annual operating test, a comprehensive written exam, or a
requalification segment evaluation. The remediation records were acceptable.). Licensee's feedback system The inspectors interviewed operator requalification instructors, training and operations
management, and two licensed operators for feedback regarding the implementation of
the licensed operator requalification program to ensure the requalification program was
meeting their needs and responsive to their recommended changes. On October 29, 2007, the inspectors conducted an in-office review of licensee
requalification exam results. These results reflected the operators' performance on the
annual operating tests; the comprehensive written exams were administered in the fall,
2006, and therefore those test results were not part of this in-office review. The
inspector assessed whether pass rates were consistent with the guidance of
- NRC [[]]
0609, Appendix I, "Operator Requalification Human Performance SDP." The inspectors
verified that: $ Crew failure rate on the dynamic simulator was less than 20 percent. (Failure rate was 0.0 percent) $ Individual failure rate on the dynamic simulator test was less than or equal to 20 percent. (Failure rate was 0.0 percent) $ Individual failure rate on the walkthrough test (job performance measures) was less than or equal to 20 percent. (Failure rate was 0.0 percent)
9$ Individual failure rate on the comprehensive written exam was less than or equal to 20 percent. (As noted above, the comprehensive written exams were administered in the fall, 2006. Test results were previously documented in NRC
IR 50-293/2006-005.) $ More than 75 percent of the individuals passed all portions of the exam. (100% of the individuals passed all portions of the exam)
b. Findings No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12)
a. Inspection Scope (2 samples) The inspectors reviewed action plans for two SSC issues and reviewed the performance
history of these
- SSC s to assess the effectiveness of Entergy=s maintenance activities. The inspectors reviewed Entergy=s
- CR s, corrective actions, and functional failure determinations made in accordance with Entergy procedures and the requirements of
ARequirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.@ In addition, the inspectors reviewed selected SSC classification, goals, corrective actions, performance criteria and monitoring plans to return the (a)(1) systems to (a)(2) status. Also, the inspectors selected a sample of
system health reports for review to evaluate the results of system performance
monitoring, material condition, and operations impact, to determine if actions taken were
reasonable and appropriate. The references used for this inspection are listed in the
to this report. The following issues were reviewed: $ Turbine Controls Subsystem failure, failed maintenance rule performance criteria of one functional failure in two years (CR-PNP-2007-03673); and $
EDG) exceeded maintenance rule performance criteria due to functional failures on October 25, 2006, and January 4, 2007 (CR-PNP-2007-0052).
b. Findings No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a. Inspection Scope (4 samples) The inspectors evaluated online and shutdown risk management for emergent and
planned activities. The inspectors reviewed maintenance risk evaluations, work
schedules, and control room logs to determine if concurrent planned and emergent
maintenance or surveillance activities adversely affected the plant risk already incurred with out-of-service components. The inspectors evaluated whether Entergy took the
10necessary steps to control work activities, minimize the probability of initiating events, and maintain the functional capability of mitigating systems. The inspectors assessed Entergy's risk management actions during plant walkdowns. Documents reviewed during
the inspection are listed in the Attachment. The inspectors reviewed the conduct and
adequacy of scheduled and emergent maintenance risk assessments for the following
maintenance and testing activities: $ Yellow risk condition during emergent unavailability of the "A"
- EDG due to an engine coolant leak in the turbo charger casing; $ Vital Motor Generator Set maintenance; $ Yellow Risk Condition during scheduled maintenance resulting in the unavailability of the
HPCI system; and $ Safety Relief Valve 3B pilot valve replacement outage shutdown risk assessment.
b. Findings No findings of significance were identified.
1R15 Operability Evaluations (71111.15)
a. Inspection Scope (5 samples) The inspectors reviewed five operability determinations associated with degraded or non-conforming conditions to determine if the operability determination was justified and if the mitigating systems or those affecting barrier integrity remained available such that no
unrecognized increase in risk had occurred. The inspectors also reviewed compensatory
measures to determine if the compensatory measures were in place and were
appropriately controlled. The inspectors reviewed licensee performance against related
UFSAR) requirements. The inspectors
reviewed the following degraded or non-conforming conditions: $
PNP-2007-04841, RHR pump P-203D revealed pump suction pressure drop outside acceptable range.
b. Findings No findings of significance were identified.
111R19 Post-Maintenance Testing (71111.19)
a. Inspection Scope (8 samples) The inspectors reviewed eight samples of post-maintenance tests (PMT) during this
inspection period. The inspectors reviewed these activities to determine whether the PMT
adequately demonstrated that the safety-related function of the equipment was satisfied,
given the scope of the work performed, and that operability of the system was restored. In addition, the inspectors evaluated the applicable test acceptance criteria to verify
consistency with the associated design and licensing bases, as well as TS requirements.
The inspectors also evaluated whether conditions adverse to quality were entered into the
corrective action program for resolution. Documents reviewed during the inspection are
listed in the Attachment. The following maintenance activities and their post-maintenance
tests were evaluated: $
- HPCI [[]]
AB@ replacement per WO 51530724. b. Findings No findings of significance were identified.
1R20 Refueling and Other Outage Activities (71111.20)
a. Inspection Scope (1 sample) The inspectors reviewed shutdown and plant restart activities associated with a planned
outage to replace the pilot on leaking Safety Relief Valve, RV-203-3B. The planned
outage commenced on December 10, 2007, and was completed on December 12, 2007.
The inspectors reviewed Entergy=s forced outage work schedule, risk evaluations, control room logs, and vessel cooldown and heatup rate data. The inspectors observed activities in the control room during the plant shutdown and startup. The inspectors conducted a
walkdown of the primary containment to verify that there was no evidence of reactor
coolant system leakage and that foreign material was being accounted for and controlled.
Documents reviewed during the inspection are listed in the Attachment.
b. Findings No findings of significance were identified.
21R22 Surveillance Testing (71111.22)
a. Inspection Scope (3 samples) The inspectors reviewed three samples of surveillance activities to determine whether the
testing adequately demonstrated equipment operational readiness and the ability to perform the intended safety-related functions. The inspectors reviewed selected
prerequisites and precautions to determine if they were met and if the tests were
performed in accordance with the procedural steps. Additionally, the inspectors evaluated
the applicable test acceptance criteria for consistency with associated design bases,
licensing bases, and TS requirements. The inspectors also evaluated whether conditions
adverse to quality were entered into the corrective action program for resolution.
Documents reviewed during the inspection are listed in the Attachment. The following
surveillance tests were evaluated: $
HPCI System Pump and Valve Quarterly and Biennial Comprehensive Operability; and $ Reactor Coolant System Leak Rate determination per TS 3/4.6.C, "Primary System Boundary Coolant Leakage."
b. Findings No findings of significance were identified. 2.
- RADIAT [[]]
- ION [[]]
- SAFE [[]]
TY Cornerstone: Occupational Radiation Safety
2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03)
a. Inspection Scope (9 samples) During the period October 15-18, 2007, the inspector conducted the following activities to
evaluate the operability and accuracy of radiation monitoring instrumentation, and the adequacy of the respiratory protection program relative to maintaining and issuing
self-contained breathing apparatus (SCBA). Implementation of these programs was
reviewed against the criteria contained in 10 CFR 20, "Standards for Protection Against
Radiation;" applicable industry standards; and Pilgrim procedures. The inspector reviewed the
- UFS [[]]
AR to identify area, process, and emergency monitors
that are installed at Pilgrim for the protection of workers. The inspectors reviewed the
current calibration records for selected instrumentation, including the Turbine Building
Radwaste Sump Area monitor (1815-8C), the Reactor Building 23' South East Access
Area monitor (1815-2D), and the Reactor Building Outside Traversing In-Core Probe
Room monitor (1815-2B). The inspector selected hand-held radiation instruments, air monitors, contamination
monitors, and electronic dosimeters currently in use in the plant, and reviewed the
13calibration records for this instrumentation. Included in this review were the calibration records for selected electronic dosimeters (DMC-2000), radiation survey instruments
(RO-2,
SAM-9), count room scalers (BC-4, SAC-4), and air samplers (H809V, Victoreen
Lapel Sampler). The inspector reviewed the maintenance records, safety interlock checks, and current
calibration source activity/dose rate determinations for the Shepard Model 78, Shepard
Model 423, and Model 773 instrument calibrators. The inspector evaluated the licensee's program for assuring quality in the radiation
monitoring instrumentation and respiratory protection programs by reviewing 16 CRs
related to radiation instrumentation,
- SC [[]]
BA's, and the monitoring of plant radiation levels to
determine if problems were identified in a timely manner and appropriate corrective
actions were taken to resolve the related issues. There were no incidents of personnel internal exposure resulting in a Committed Effective
Dose Equivalent > 50 mrem that would require an in-depth evaluation of whole body
counting instrumentation and bioassay techniques. The inspector reviewed actions for radiation worker and radiation protection technician
errors to determine whether the corrective actions were adequate to prevent recurrence. The inspector verified calibration due dates and observed a technician performing source
checks on a variety of instruments including portable radiation survey instruments (RO-2,
Wide Range Telepole), contamination survey instruments (RM-14s, SAM 9), count room
scalers (BC-4), and personal contamination monitors (PPM-1,
SCBAs staged for use in the control
room, Radiological Controlled Area access location, and the fire brigade equipment
staging area in the fire service pump building. The inspector observed a technician
perform an inspection of six of the ten units staged for use. The inspector observed a
technician fill two
- SC [[]]
BA air bottles from the air compressor unit. The sample results for
breathing air, used to refill the
CGA-G-7.1-2004 Grade D standards. The inspector evaluated the adequacy of the respiratory protection program regarding the
issuance of
- SC [[]]
BAs to workers. Training and qualification records for licensed operators,
radiation protection technicians, and fire brigade members required to wear
- SC [[]]
BA's, in
the event of an emergency, were reviewed.
b. Findings No findings of significance were identified.
142PS3 Radiological Environmental Monitoring Program and Radioactive Material Control Program (71122.03)
a. Inspection Scope (1 sample) During the period October 15-18, 2007, the inspector conducted the following activity to
determine whether the licensee's surveys and controls are adequate to prevent the
inadvertent release of licensed materials into the public domain. Implementation of these
controls was reviewed against the criteria contained in 10 CFR 20, "Standards for
Protection Against Radiation;" TS; and Entergy procedures. This inspection activity
represents completion of one sample relative to this inspection area. The inspector observed the radioactive material survey and release locations. The
methods used for control, survey, and release were inspected and included observations
of the performance of personnel surveying and releasing material for unrestricted use and
verifying that the work is performed in accordance with plant procedures.
b. Findings No findings of significance were identified. 4.
- OTHER [[]]
- ACTIVI [[TIES [OA]]]
PI) (71151)
.1 Mitigating System Cornerstone
a. Inspection Scope (2 samples) The inspectors sampled data for the Mitigating System Performance Index
RBCCW) for the 4th quarter
2006 and 1st, 2nd and 3rd quarter 2007 to assess the completeness and accuracy of the
reported information. The inspectors reviewed operator logs, CRs, maintenance rule
documents, maintenance records, Licensee Event Reports (LERs), system health reports,
and plant process computer information. The acceptance criteria used for the review
were Nuclear Energy Institute (NEI) 99-02, Revision 5, "Regulatory Assessment
Performance Indicator Guidelines."
b. Findings No findings of significance were identified.
.2 Physical Protection Cornerstone
a. Inspection Scope (3 samples) The inspectors performed a review of PI data submitted by the licensee for the Physical
Protection Cornerstone. The review was conducted of the licensee=s programs for gathering, processing, evaluating, and submitting data for the Fitness-for-Duty, Personnel
15Screening, and Protected Area Security Equipment
PIs had been properly reported as specified in NEI 99-02. The review
included the licensee=s tracking and trending reports, personnel interviews, and security event reports for the
- PI data collected since the last security baseline inspection. The inspector noted from the licensee=s submittal that there were no reported failures to properly implement the requirements of 10
CFR 73, "Physical Protection of Plants and Materials," and 10 CFR 26, "Fitness for Duty Programs," during the reporting period. This
inspection activity represents the completion of three samples relative to this inspection
area; completing the annual inspection requirement.
b. Findings No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
.1 Review of Items Entered into the Corrective Action Program (CAP)
a. Inspection Scope The inspectors performed a screening of each item entered into the licensee's CAP. This
review was accomplished by reviewing printouts of each CR, attending daily screening
meetings and/or accessing the licensee's database. The purpose of this review was to
identify conditions such as repetitive equipment failures or human performance issues
that might warrant additional follow-up.
b. Findings No findings of significance were identified.
.2 Semi-Annual Review to Identify Trends
a. Inspection Scope (1 sample) As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"
the inspectors performed a review of Entergy=s CAP and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors review was focused on repetitive equipment and corrective maintenance issues but also
considered the results of daily inspector
CAP trend reports and site CAP
performance indicator data. The inspectors review considered the six month period of
June through December, 2007, although the inspectors also evaluated the trend review
results discussed in
- NRC [[]]
IR 05000293/2007003, which reviewed CRs from October 2006
through May 2007. Documents reviewed during the inspection are listed in the
Attachment.
16b. Assessment and Observations No findings of significance were identified. The inspectors noted a number of plant
equipment configuration control issues discussed in the third quarter 2007 Pilgrim Station
Quarterly Trend Report, including: $
PNP-2007-02651, EDG failed to start (likely due to fuel rack and governor left in full fuel position). The report concluded that the number of issues Adoes not exhibit an adverse or emerging trend,@ but that Operations Management considers the number of Amispositionings@ to be at an unacceptable level. The inspectors also considered the number of issues discussed in the report to be at an unacceptable level, however, the inspectors also concluded that
these issues represent a low level trend in the area of configuration control. The
inspectors have discussed this trend with licensee management and will continue to monitor configuration control issues at Pilgrim during this assessment period.
.3 Annual Sample: Review of Outage CRs
a. Inspection Scope (1 sample) The inspectors reviewed a sample of CRs from Pilgrim's 2007 refueling outage to
determine whether CRs initiated during the outage were processed and closed in
accordance with Pilgrim procedures. The inspectors reviewed two Apparent Cause
Evaluations conducted by Pilgrim. The inspectors evaluated whether corrective actions
taken by Pilgrim addressed each CR as well as the overall process. Documents reviewed
are listed in the Attachment.
b. Assessment and Observations No findings of significance were identified. The inspectors determined that there were
many instances where the condition review group (CRG) closed a lower level (Category
D)
CR
with a general statement such as "Corrective actions for the CR were reviewed by the
responsible manager. Upon the manager's recommendation, this CR is being closed."
This practice resulted in a condition where corrective actions for a particular issue could
not be tracked or demonstrated. Pilgrim has since discontinued this practice as an
acceptable closure strategy for Category D CRs.
17.4 Annual Sample: Review of Motor Operated Valve (MOV) Hydraulic Lock
a. Inspection Scope (1 sample) The inspectors selected
PNP-2006-04328 for detailed review. The CR was written to
determine the cause of a safety-related
RHR system during routine
surveillance testing. The inspectors reviewed the licensee's root cause analysis,
corrective actions, and the prioritization of the corrective actions.
b. Assessment and Observations No findings of significance were identified. The inspectors determined that the licensee
performed a thorough root cause analysis and took timely corrective actions to prevent
recurrence. The root cause was determined to be hydraulic locking of the MOV actuator
due to grease found inside of the spring package. The grease prevented the spring
package from compressing which in turn prevented the thermal overloads from tripping.
The tripping of the thermal overloads stops the motor and provides the indication that the
valve is closed. The root cause analysis determined that newer MOVs in the plant were not susceptible to
hydraulic lock because the valves have an internal grease relief path from the spring
package to the actuator housing. However, most MOVs at Pilgrim did not have the
internal grease relief path. Immediate corrective actions included looking inside the
spring package of all safety-related MOVs for grease. Long term corrective actions for
this issue included a design modification to provide an external grease relief path from the
spring package back to the actuator housing. All of the high priority valves have been
modified. The last low priority valve to receive this modification is scheduled to be
performed in the next refueling outage. The inspectors determined that the prioritization
of the corrective actions was appropriate.
.5 Annual Sample: Follow-up Review of Component Design Bases Inspection (CDBI) Finding Regarding the Inadequate Operability Determination for the
- HPCI Turbine Trip Solenoid Failure a. Inspection Scope (1 sample) The inspectors reviewed the corrective actions for a finding identified during the
CDBI and
documented in inspection report number 05000293/2006006. The finding was associated
with Entergy=s failure to declare the
PNP-2006-01460 to determine whether the corrective actions were appropriate and completed. As part of this review, the inspectors
examined various safety system operating procedure changes to assess their adequacy.
The documents reviewed are listed in the Attachment to this report.
b. Assessment and Observations
No findings of significance were identified. Entergy=s initial failure to declare
- HPCI inoperable was due to licensing and operations department management focusing on the ability of the
HPCI system to perform its accident analysis function versus a discussion of
18the
3.2.B, "Protective Instrumentation Core and Containment Cooling Systems - Initiation and
Control." As a result, the
- HP [[]]
CI system should have been considered inoperable
regardless of the ability of the system to perform its accident analysis function. The inspectors determined that the licensee=s corrective actions were appropriate. Entergy determined the failure to declare
- HP [[]]
CI inoperable was due to a lack of independence of the operations department and licensing departments in reviewing operability determinations. The inspectors noted that Entergy immediately implemented operations
department training regarding independent review of emerging TS issues. Also, Entergy
revised safety system operating procedures to include a section on TS instrumentation
requirements.
4OA3 Event Follow-up (71153)
.1 Infrequently Performed Evolution: MG Set Power Transfer
a. Inspection Scope (1 sample) On October 3, 2007, Pilgrim operators performed a planned manual transfer of vital
alternating current (AC) power from its normal power source, the vital MG set, to its
alternate power source, bus B15, with the plant at power. This infrequently performed
evolution was conducted to remove the vital MG set from service for repairs. The evolution
posed several challenges to Pilgrim operators because the transfer of the vital AC power
from its normal to its alternate source would cause a momentary interruption in vital AC
power. Similar evolutions in the past had resulted in complications such as the receipt of
reactor building isolation signals, feed regulating valve position lock ups, and recirculation
pump scoop tube position lock ups. Entergy developed a new procedure for this evolution,
Procedure 2.2.16, Attachment 8, "A Manual Transfer of Y2 to Motor Control Center (MCC)
B15 with the Units On-line." The procedure established several compensatory measures
to mitigate the effects of a component malfunction or unexpected response. For instance,
operators were briefed on Procedure 2.4.49, Section 4.4, "A Manual Lockup of Feed
Regulating Valve(s) from the Condenser Bay," and were stationed outside the condenser
bay to take manual control of the valves if needed. Additionally, operators inserted a
reactor building isolation signal before the vital power transfer, to prevent the signal from
coming in during the transfer. The inspectors reviewed the procedure and observed the
evolution from the control room to assess operator actions, command and control, and the
adequacy of communications within the control room and between the control room and
the field.
b. Findings No findings of significance were identified.
19.2
LER 05000293/2007-005-00, Reactor Scram Resulting from Low Vacuum Turbine Trip
a. Inspection Scope The inspectors reviewed Entergy=s actions associated with
- LER 50-293/2007-05-00, which discussed the July 10, 2007, low vacuum turbine trip and automatic reactor scram event. The inspectors reviewed the licensee=s
LER and associated root cause evaluation. Additionally, the inspectors verified that follow-up actions, taken or planned, were appropriate to address the event. This LER is closed.
b. Findings Introduction: A Green self-revealing finding was identified for Entergy=s failure to ensure the proper verification and calibration of vacuum trip switch
RFO) 16. Specifically, personnel did not ensure that the proper verification/calibration
technique was employed to determine the as-found low condenser vacuum turbine trip
setpoint. Additionally, when the technician identified that the as-found data was
significantly outside of historical as-found values, he did not question the validity of the data
nor did he obtain a peer check. The technician then calibrated the instrument using the
incorrect as-found data which resulted in an incorrect low vacuum trip setpoint and a
subsequent turbine trip and reactor scram on July 10, 2007. Description: On July 10, 2007, an unplanned automatic reactor scram occurred while performing condenser thermal backwashes at approximately 48 percent power. The reactor
protection system (RPS) scram signal was initiated by the trip of the main turbine on low
condenser vacuum. Pilgrim operators stabilized the plant in a shutdown condition and made a four-hour notification to the NRC. Post scram review of the as-found setpoint for
vacuum trip switch, VTS-1, revealed that the trip setpoint was set to actuate at 24.35@ Hg rather than the expected 21.95@ B 22.45@ Hg. Entergy recalibrated the vacuum switch and restored the plant to 100 percent power on July 16, 2007.
Entergy conducted a root cause evaluation of the unplanned scram and summarized their results in LER 2007-005-00, "Reactor Scram Resulting from Low Vacuum Turbine Trip."
Entergy determined that the root cause of the event was that the technician who had
calibrated the
RFO 16 had not properly implemented human
performance tools (e.g., training) for this particular type of large volume instrument to
ensure a proper calibration. Specifically, since the bellows for VTS-1 are very large, the
vacuum must be decreased slowly during the calibration in order for an accurate setpoint to
be obtained. While obtaining the as-found setpoint, the technician did not decrease the
vacuum slowly which resulted in faulty as-found results. Additionally, when the as-found
data suggested that the vacuum switch was considerably outside of historical results, the
technician did not question the validity of the data nor did he obtain a peer check. The
technician then made adjustments to the instrument using the incorrect as-found data.
20Entergy=s root cause report also discussed several weaknesses with Procedure
ATurbine Generator and Auxiliary Instruments Calibration.@ Specifically, the root cause report noted that Aadditional details in the procedure would provide an additional barrier to ensure the proper calibration technique is achieved.@ However, the inspectors noted that Entergy had not identified these procedural weaknesses as a contributing cause to this event. The inspectors concluded that the lack of procedural specificity and guidance
contributed to the improper calibration of VTS-1. Entergy=s corrective actions for this aspect included adding steps to the procedure to decrease the vacuum at a slower rate, to include detailed guidance on the adjustments of the trip and span of the vacuum trip
assembly, and to require supervisory review of as-found data and testing techniques prior
to performing adjustments.
Analysis: The performance deficiency associated with this finding is that Entergy did not ensure the proper verification and calibration of vacuum trip switch
RFO 16.
The improper setpoint resulted in a low vacuum turbine trip and consequent automatic
reactor scram on July 10, 2007. This finding is more than minor because it is associated
with the human performance attribute of the Initiating Events Cornerstone and affects the
cornerstone objective of limiting the likelihood of those events that upset plant stability during power operations. The inspectors conducted a Phase 1 screening in accordance
with IMC 0609, "Significance Determination Process," Appendix A, "Reactor Inspection
Findings for At-Power Situations." The finding was determined to be of very low safety
significance (Green) because it did not contribute to both the likelihood of a reactor trip and
the likelihood that mitigation equipment would be unavailable. This finding has a
cross-cutting aspect in the area of Human Performance, Work Practices, because Entergy
proceeded in the face of uncertainty or unexpected circumstances by continuing with the
calibration procedure even though the vacuum trip switch setpoint was found significantly
outside of historical as-found values. H.4(a) Enforcement: Enforcement action does not apply because the performance deficiency did not involve a violation of a regulatory requirement in that the vacuum trip switch is not a
safety-related component. Entergy has entered this issue into their corrective action
program as
PNP-2007-3231. Corrective actions included recalibrating VTS-1 before
the plant restart, providing remedial training for the technician who had conducted the
improper calibration, and adding vacuum switch fundamentals as a continuing training
topic for the instrumentation and controls (I&C) technicians. Additional corrective actions
planned by Entergy include revising Procedure 8.F.51 to include more detailed guidance
and to require a supervisory review of as-found data prior to performing adjustments;
conducting just-in-time training prior to the RFO 17 vacuum trip switch setpoint verification
and calibration; and identifying and revising other I&C procedures involving critical
calibrations. Because this violation does not involve a violation of regulatory requirements and has a very low safety significance, it is identified as
- FIN 05000293/2007005-01, Improper Calibration of Vacuum Trip Switch Results in an Automatic Reactor Scram. 4
OA6 Meetings, Including Exit On October 18, 2007, an Occupational Radiation and Public Radiation Safety exit meeting
was conducted. The preliminary inspection results were presented to Robert Smith,
21General Manager Pilgrim Operations, and other members of the Pilgrim staff. The licensee did not identify any material as proprietary during this inspection. On October 18, 2007, the Security inspection results were presented to members of
licensee management. On January 9, 2008, the resident inspectors conducted an exit meeting and presented the
preliminary inspection results to Mr. Kevin Bronson, Site Vice President, and other
members of the Pilgrim staff. The inspectors confirmed that no proprietary information was
provided or examined during the inspection.
- ATTACH [[]]
- MENT [[:]]
- SUPPLE [[]]
- MENTAL [[]]
- INFORM [[]]
- SUPPLE [[]]
- MENTAL [[]]
- INFORM [[]]
- ATION [[]]
- KEY [[]]
- POINTS [[]]
- OF [[]]
CONTACT
Licensee personnel:
S. Bethay Nuclear Safety Assurance Director
K. Bronson Site Vice President, Pilgrim
H. Bouska Supervisor, Operations Training
D. Burke Security Manager
L. Foreaker Supervisor, Radiation Instrumentation
J. Henderson Manager, Radiation Protection
M. Gakka Licensing
T. Kelly Technician, Radiation Protection
R. Larson Technician, Radiation Protection
W. Lobo Licensing Engineer
J. Lynch Licensing Manager
F. Marcussen Protective Services Department Manager
C. McMorrow Senior Operations Instructor
D. Noyes Operations Director
M. Santiago Superintendent, Nuclear Training
L. Seehaus Technician, Radiation Protection
R. Smith Plant Operations General Manager
- LIST [[]]
- OF [[]]
- ITEMS [[]]
- CLOSED [[]]
- AND [[]]
- DISCUS [[]]
SED
Opened and Closed
05000293/2007005-01 FIN Improper Calibration of Vacuum Trip Switch Results in an Automatic Reactor Scram
Closed
05000293/2007-005-00
- LIST [[]]
- OF [[]]
- DOCUME [[]]
- NTS [[]]
- REVIEW [[]]
- UFS [[]]
- NRC [[]]
- NRC [[]]
IN 98-002, Nuclear Power Plant Cold Weather Problems and Protective Measures
Procedure 8.C.40, Seasonal Weather Surveillance, Attachment 1, Cold Weather Preparations, Revision 19 Procedure 2.2.35, Condensate Storage and Transfer System, Revision 40
A-2 Section 1R04 Drawing M215 Sheet 2, Revision 48,
- P& [[]]
ID Cooling Water System Reactor Building
Drawing M215 Sheet 5, Revision E8, Composite
- P& [[]]
ID Cooling Water System Reactor Building
Procedure 2.2.30, Revision 65,
- RBC [[]]
Procedure 2.2.19, Residual Heat Removal System, Revision 95
M241, P21D, Residual Heat Removal System, Revision 47
- PN [[]]
PS Procedure 2.2.21, Revision 72, High Pressure Coolant Injection System
Procedure 7.1.44, "Sampling of Charcoal Cells in
- SBGT and Control Room Environmental Filters' Systems for Methyl Iodide Testing", completed on 11/28/06 for "B"
- SBGT [[]]
- SBGT [[]]
- PN [[]]
- PNPS Final Safety Analysis Report, Revision 10, Chapter 7.18, Reactor Building Isolation and Control System
Pilgrim
SBGT System and Control Room High Efficiency Air Filtration System
Procedure 2.2.30,
- RBC [[]]
CW System, Revision 65
Procedure
AB@ Train Functional Test of Individual Load Shed Components, Revision 18 Section 1R05 Pre Fire Plan, Screenhouse Building EL. 23'
Pre Fire Plan, Reactor Building Quads,
ER-Q, Updated Fire Hazards Analysis, Revision E5
Procedure 5.5.2, Special Fire Procedure, Revisions 29 and 37
- PNPS Procedure 8.B.17.2, Inspection of Fire Damper Assemblies, Attachment 1, Revision 9, completed 4/3/07
PNPS Procedure 8.B.17.2, Inspection of Fire Damper Assemblies, Attachment 11, Revision 9, completed 4/4/07
Section 1R06
- NRC [[]]
IN 2007-01, Recent Operating Experience Concerning Hydrostatic Barriers
Procedure
- PNP -03457 Section 1R11 Lesson Plan O-RO-07-02-01, Revision 4, Emergency Classification and Notification
- NRC [[]]
EP-IP-100, Revision 26, Emergency Classification and Notification
A-3EP-IP-300, Revision 6, Offsite Radiological Dose Assessment
Lesson Plan O-RO-07-03-03, Revision 0,
- PNP -2007-4587, Control Room does not have the same weather assessment capability (160' Met Tower) for
- EOE [[]]
- IP [[-400 states that core temperature >2400F is indication of substantial core damage, this temperature is not able to be obtained Section 1R12]]
- EDG [[(a)(1) Action Plan Health Report, System 02, Reactor Recirculation 3rd Qtr 2007 Health Report, System 29, Salt Service Water 3rd Qtr 2007 Health Report, System 01, Main Steam, 3rd Qtr 2007 10/09/2007, MR Expert Panel Meeting Minutes Section 1R13 Risk Management Actions]]
- EDG turbo charger gas inlet casing Procedure 2.2.16, Revision 50, Attachment 8, Manual Transfer of Y2 to
HPCI System
Equipment out of service (EOOS) quantitative risk assessment tool
Procedure 3.M.1-45, Outage Shutdown Risk Assessment, Revision 6
Risk Assessment Review Checklist for 12/10 08:00 to 12/12 18:00
- PNPS for 12/10 0:00 to 12/13 12:00 Risk Assessment Review Checklist for 12/10 08:00 to 12/12 18:00, Revision A Section 1R15
MPR Setpoint Adjustments
Apparent Cause Evaluation for
- HP [[]]
CI System
TS 3.12, Fire Protection, Alternate Shutdown Panels
A-4CR-PNP-2007-04724,
HPCI Safety Function
Entergy procedure
- CR -PNP-2007-04841, Initial operability review for pump P203D pump suction pressure drop value not acceptable
- AB @ Operability-Pump Quarterly and Biennial (Comprehensive) Flow Rate Tests and Valve Tests 51535468 01, Work Order,
- RHR Inservice pump test data sheets for 11/26 and 12/3/2007 Section 1R19 Procedure 8.5.3.2.1, Revision 19, Attachment 1D, Quarterly and Biennial (Tech Spec/IST) Test Procedure for
CR 2007-04264
Apparent Cause Evaluation for CR 2007-4274
M1J18-11, Elementary Diagram High Pressure Coolant Injection System
4533K40-800, page 43/44, Figure 24: Schematic Diagram of Type 540-01 and 540-51 controller (for
- HPCI [[]]
HPCI MO-14
Procedure 1.3.34, Operations Administrative Policies and Processes, Revision 113
Procedure 2.2.21.5,
- HP [[]]
CI Injection and Pressure Control, Revision 13
Procedure 8.5.4.1,
- HPCI System Pump and Valve Quarterly and Biennial Comprehensive Operability, Revision 102 Procedure 8.5.4.4,
HPCI System Instrument Calibration, Revision 65
Procedure
HPCI Injection Flow Controller
50.72 Event Report to
- USN [[]]
- RC High Pressure Coolant Injection Inoperable, dated November 20, 2007 Control Room (day) Shift Narrative Logs, dated November 19, 2007
- PNP -2007-04937, Air leakage identified at connection between the solenoid valve and the manifold Procedure
- 3.M. 4-6, Removal, installation, Test, Disassembly, Inspection, and Reassembly of Main Steam Relief Valves 3379-270-3 E5, Main Steam
SRV Sheets 1, 2, 3 and 4
A-53379-271-1 E1, Main Steam
- WO [[]]
WO Automatic Depressurization System subsystem manual opening of relief valves Procedure 2.1.19, Suppression chamber temperatures
Procedure 8.5.6.2, Special test for ADS system manual opening of relief valves
Section 1R20 PNP On-Line Master Schedule, dated 11/30/07, 12/10/07, and 12/11/07
Procedure 2.1.5, Controlled Shutdown from Power, Revision 103
Procedure 2.2.19.1, Residual Heat Removal System - Shutdown Cooling Mode of Operation, Revision 24 Procedure 2.1.1, Startup from Shutdown, Revision 162
Procedure, 2.1.7, Vessel Heatup and Cooldown, Revision 52, completed 12/12/2007
Section 1R22 Procedure 8.5.5.1, Revision 56,
Procedure 6.1-220, Radiological Controls for High Risk Evolutions, Revision 2
Procedure
RP-131, Attachment 9.2, Revision 3, Air Sampling results from November 19, 2007
Control Room (day) Shift Narrative Logs, dated 11/20/2007
Technical Specification 3.5.C, High Pressure Coolant Injection
- UFS [[]]
PNPS - Entergy Relief Request PR-03 High Pressure Coolant Injection Pump, dated August 29, 2005 Procedure 2.1.15, Daily Surveillance Log, Revision Procedure 8.M.2-5, Drywell Drain Sump Integrator, Revision 9, Attachment 1, completed 10/18/07
Procedure 8.M.2-5, Drywell Drain Sump Integrator, Revision 9, Attachment 2, completed 10/6/05
Drawing C-75, Reactor Building Foundations Drywell Concrete @ El. 9'-2, Revision 4
ER# 06110910, Attachment 9.1
Control Room Shift Narrative Logs, dated 12/5/2007 through 12/7/2007
Sections
OS2/20S3 6.5-003, Revision 8, Radiation Protection Instrumentation Calibration Frequency
6.5-160, Revision 31, Calibration of the Area Radiation Monitoring System
6.5-170, Revision 21, Calibration of Ventilation System Radiation Monitors Using
RO-2/RO2A or RO-20 Ion Chamber
6.5-311, Revision 10, Calibration of the Eberline Model RO-7 Radiation Monitor
6.5-341, Revision 11, Calibration of the MDC 2000S Electronic Dosimeter
6.7.1-106, Revision 14, Inspection and Testing of Respiratory Protection Equipment
6.7.1-201, Revision 8, Operation of the
- SC [[]]
BA Air Compressor
A-6EN-RP-121, Revision 1, Radioactive Material Control
RP-502, Revision 1, Inspection and Maintenance of Respiratory Protection Equipment
Calibration Records:
Electronic Dosimeter Calibration (Serial Nos. 176631, 219267, 178032, 177025, 170628)
E-520 (Serial No. 722)
SAC-4 (Serial No. 1402)
BC-4 (Serial No. 484)
Victoreen Lapel Sampler (Serial No. c1138)
H809V (Serial No. 6168)
PM-7(Serial No. 600, 392)
Wide Range Telepole (Serial No. 6603-027)
RO-2 (Serial No. 3410)
RO-2A (Serial No. 3295)
RO-20 (Serial No. 325, 285)
RO-7 (Serial No. 1030)
RM-14 (Serial No. 8565)
SAM-9 (Serial No. 308)
- SC [[]]
BA Numbers :1, 2, 3, 4, 5, 10, 11, 12, 13, 14
Miscellaneous Records & Reports:
Mask Qualification List
Root Cause Analysis Report for
Instructional Module C-FB-02-02-01, Revision 7 Self-Contained Breathing Apparatus
Section 4OA2 Limitorque Maintenance Update 90-1
Limitorque Maintenance Update 88-2
DRN 07-01007, Limitorque Valve Controls
Third quarter 2007 Pilgrim Station Quarterly Trend Report
- NRC [[]]
PNP-2007-02651, EDG failed to start (likely due to fuel rack and governor left in full fuel position)
A-7Procedure 2.2.21, High Pressure Coolant Injection System, Revision 72 Procedure 2.2.19, Residual Heat Removal System, Revision 95
Procedure 2.2.3, Automatic Depressurization System, Revision 23
Procedure 2.2.8, Emergency Diesel Generator, Revision 90
- LIST [[]]
- OF [[]]
- ACRONY [[]]
- ADA [[]]
MS Agencywide Documents Access and Management System
- CD [[]]
BI component design bases inspection
CFR Code of Federal Regulations
CR condition report
CRG Condition Review Group
DRP Division of Reactor Projects
DRS Division of Reactor Safety
EAL emergency action level
EDG emergency diesel generator
gpm gallon per minute
Hg mercury
- HP [[]]
CI high pressure coolant injection
I&C instrumentation and controls
IMC Inspection Manual Chapter
IR Inspection Report
LER Licensee Event Report
MCC motor control center
MG motor generator
MO motor-operated
MOV motor-operated valve
MPR mechanical pressure regulator
mrem millirem
NEI Nuclear Energy Institute
NRC Nuclear Regulatory Commission
- PA [[]]
RS Publicly Available Records
PI Performance Indicator
- PN [[]]
- RBC [[]]
- RC [[]]
IC reactor core isolation cooling
RFO refueling outage
RV relief valve
- SB [[]]
- SC [[]]
BA self-contained breathing apparatus
A-8SDP Significance Determination Process SRM source range monitor
SRO senior reactor operator
SSC system, structure, or component
SSW salt service water
- UFS [[]]
AR Updated Final Safety Analysis Report
URI unresolved item
VTS vacuum trip switch